crp/OpenMC
1
0
Fork 0
mirror of https://github.com/openmc-dev/openmc.git synced 2026-07-29 06:35:48 -04:00
Code Issues Projects Releases Packages Wiki Activity Actions 5

added back in MGXSLibrary.from_hdf5 (for some reason it disappeared) and incorporated conversion of scatter matrix to vector for mgxs plotting

Browse source
This commit is contained in:
Adam Nelson 2016-11-19 06:21:39 -05:00
parent ea8181f2b6
commit be282ed7dd
3 changed files with 71 additions and 8 deletions
Download patch file Download diff file Expand all files Collapse all files

33
openmc/plotter.py
View file

@ -10,10 +10,14 @@ import openmc.data
PLOT_TYPES = ['total', 'scatter', 'elastic', 'inelastic', 'fission',
'absorption', 'capture', 'nu-fission', 'nu-scatter', 'unity',
'slowing-down power', 'damage']
# Supported keywoards for multi-group cross section plotting
PLOT_TYPES_MGXS = ['total', 'absorption', 'fission', 'kappa-fission',
'chi', 'chi-prompt', 'nu-fission', 'prompt-nu-fission',
'inverse-velocity', 'unity']
PLOT_TYPES_MGXS = ['total', 'absorption', 'scatter', 'fission',
'kappa-fission', 'chi', 'chi-prompt', 'nu-fission',
'prompt-nu-fission', 'inverse-velocity', 'unity']
# Add on values for scattering moments
PLOT_TYPES_MGXS += ['scatter-' + str(i)
for i in range(0, openmc.mgxs.MAX_LEGENDRE + 1)]
# Special MT values
UNITY_MT = -1
@ -822,6 +826,21 @@ def _calculate_mgxs_nuc_macro(this, types, library, temperature=294.):
for i, line in enumerate(types):
if line == 'unity':
data[i, :] = 1.
elif line.startswith('scatter'):
# We have to remove the outgoing dependence
attr = line.replace(' ', '_').replace('-', '_')
matrix = xsdata.scatter_matrix[t]
# Sum over outgoing groups
vector = np.sum(matrix, axis=1)
# Now get the actual order of interest
if line == 'scatter':
order = 0
else:
order = int(line.split('-')[1])
if order < xsdata.xs_shapes["[G][G'][Order]"][-1]:
data[i, :] = vector[:, order]
else:
data[i, :] = 0.
else:
attr = line.replace(' ', '_').replace('-', '_')
data[i, :] = getattr(xsdata, attr)[t]
@ -883,10 +902,10 @@ def _calculate_mgxs_elem_mat(this, types, library, temperature=294.,
nuclides = {this._macroscopic: (this._macroscopic, this.density)}
else:
# Expand elements in to nuclides with atomic densities
nuclides = this.get_nuclide_atom_densities(ce_cross_sections)
nuclides = this.get_nuclide_atom_densities()
# For ease of processing split out nuc and nuc_density
nuc_multiplier = [nuclide[1][1] for nuclide in nuclides.items()]
nuc_fraction = [nuclide[1][1] for nuclide in nuclides.items()]
else:
T = temperature
# Expand elements in to nuclides with atomic densities
@ -894,7 +913,7 @@ def _calculate_mgxs_elem_mat(this, types, library, temperature=294.,
cross_sections=ce_cross_sections)
# For ease of processing split out nuc and nuc_fractions
nuc_multiplier = [nuclide[1] for nuclide in nuclides]
nuc_fraction = [nuclide[1] for nuclide in nuclides]
nuc_data = []
for nuclide in nuclides.items():
@ -908,6 +927,6 @@ def _calculate_mgxs_elem_mat(this, types, library, temperature=294.,
data[line, :] = 1.
else:
for n in range(len(nuclides)):
data[line, :] += nuc_multiplier[n] * nuc_data[n][line, :]
data[line, :] += nuc_fraction[n] * nuc_data[n][line, :]
return data